ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 48

DOE Review Release 08/31/2012

Document Title:

Paducah Gaseous Diffusion Plant –

Occupational External Dose

Document Number: ORAUT-TKBS-0019-6

Revision: 04

Effective Date: 08/24/2012

Type of Document: TBD

Supersedes: Revision 03, ORAUTOTIB-0031 Rev 00 PC-2

and ORAUT-OTIB-0017 

Rev 01

Subject Expert(s): Jodie L. Phillips

Approval: Signature on File Approval Date: 08/21/2012

Jodie L. Phillips, Document Owner

Concurrence: Signature on File Concurrence Date: 08/21/2012

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 08/21/2012

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 08/21/2012

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 08/24/2012

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 2 of 48

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

08/24/2004 00 New technical basis document for the Paducah Gaseous Diffusion 

Plant – Occupational External Dose. First approved issue. Initiated 

by Jay J. Maisler.

03/29/2005 01 Approved issue of Revision 01. Initiated by Jay J. Maisler.

08/30/2006 02 Revision initiated to address comments from earlier unresolved 

comments received on TBD. Approved issue of Revision 02. 

Incorporates additional information obtained through additional data 

capture for the Paducah Gaseous Diffusion Plant. Incorporates 

revised standard language into Purpose section and adds a Scope 

section. This revision addresses comments from the Worker 

Outreach meeting with the United Steelworkers Local 5-550 and 

Security, Police, Fire Professionals of America Local 111 held on 

February 10, 2005. Constitutes a total rewrite of the document. This 

revision results in a reduction in assigned dose and no PER is 

required. Training required: As determined by the Task Manager. 

Initiated by Paul A. Szalinski.

04/04/2007 03 Approved Revision 03 initiated to incorporate Attributions and 

Annotations section. Constitutes a total rewrite of the document. 

This revision results in no change to the assigned dose and no PER 

is required. No further changes occurred as a result of formal 

internal review. Training required: As determined by the Task 

Manager. Initiated by Daniel S. Mantooth.

08/24/2012 04 This document was predominantly revised to address SC&A 

concerns about the document as identified in SC&A-TR-TASK1-

0016, Review of the NIOSH Site Profile for the Paducah Gaseous 

Diffusion Plant. Wording was revised throughout the document to

add clarity. Acronyms and Abbreviations, References, and Glossary

Sections have been updated. Table 6-1 has been revised to change 

four-element film to two-element film from 1953 through July 1, 1960, 

and to add an annual badge exchange option beginning in 1989. 

Building start dates in Table 6-3 have been clarified. Guidance for 

assigning neutron dose has been revised. Guidance for 99Tc 

exposure has been added as Attachment A. Guidance for coworker 

dose assignment based on ORAUT-OTIB-0031 has been 

incorporated into this TBD as Attachment B. The application of 

Construction Trade Worker doses has been added. Guidance for 

skin dose assignment based on ORAUT-OTIB-0017 has been 

incorporated into this TBD as Attachment C. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Jodie L. Phillips.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 3 of 48

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations..................................................................................................................5

6.1 Introduction .................................................................................................................................7

6.1.1 Purpose...........................................................................................................................8

6.1.2 Scope..............................................................................................................................8

6.2 Basis of Comparison...................................................................................................................8

6.3 Dose Reconstruction Parameters ...............................................................................................9

6.3.1 Administrative Practices..................................................................................................9

6.3.2 Dosimetry Technology ..................................................................................................12

6.3.2.1 Beta/Photon Dosimeters ................................................................................12

6.3.2.2 Neutron Dosimeters .......................................................................................13

6.3.3 Calibration.....................................................................................................................13

6.3.3.1 Beta/Photon Dosimeters ................................................................................13

6.3.3.2 Neutron Dosimeters .......................................................................................14

6.3.4 Workplace Radiation Fields ..........................................................................................15

6.3.4.1 Beta/Photon Fields .........................................................................................15

6.3.4.2 Workplace Beta/Photon Dosimeter Response ...............................................17

6.3.4.3 Neutron Fields ................................................................................................17

6.3.4.4 Workplace Neutron Dosimeter Response ......................................................18

6.4 Adjustments to Recorded Dose ................................................................................................18

6.4.1 Photon Dose .................................................................................................................18

6.4.2 Nonpenetrating Dose ....................................................................................................19

6.4.3 Neutron Dose................................................................................................................19

6.5 Missed Dose .............................................................................................................................19

6.5.1 Estimating Missed and Unmonitored Photon Deep Dose .............................................19

6.5.1.1 Construction Trade Workers...........................................................................20

6.5.2 Estimating Missed and Unmonitored Shallow Dose .....................................................20

6.5.3 Estimating Missed and Unmonitored Neutron Dose .....................................................21

6.6 Uncertainty................................................................................................................................22

6.7 Dose Reconstruction.................................................................................................................22

6.8 Organ Dose...............................................................................................................................23

6.9 Attributions and Annotations .....................................................................................................23

References ...........................................................................................................................................24

Glossary................................................................................................................................................28
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 4 of 48

ATTACHMENT A, EXTERNAL EXPOSURE TO TECHNETIUM-99 AT THE PADUCAH 

GASEOUS DIFFUSION PLANT.............................................................................36

ATTACHMENT B, PGDP COWORKER DOSE ASSIGNMENT...........................................................40

ATTACHMENT C, SKIN DOSE ASSIGNMENT FOR PGDP ...............................................................47

LIST OF TABLES

TABLE TITLE PAGE

6-1 Dosimeter type, period of use, exchange frequency, MDL, and potential 

annual missed dose (rem) ........................................................................................................11

6-2 Major radiation sources.............................................................................................................15

6-3 Buildings and periods of operation............................................................................................16

6-4 Dose fractions for PORTS HEU storage vault in Building 345..................................................18

6-5 Missed neutron dose assignment .............................................................................................22

A-1 PGDP facilities with 99Tc exposure potential.............................................................................37

A-2 Job titles for workers with possible 99Tc exposure ....................................................................37

B-1 Missed external doses based on the body of this TBD and Attachment C ...............................42

B-2 Annual PGDP external coworker doses modified to account for missed dose .........................45

B-3 Annual PGDP external penetrating coworker doses modified in accordance 

with ORAUT-OTIB-0052 ...........................................................................................................45

C-1 Examples of skin dose assignments for PGDP badge readings in 1970..................................48

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Estimated dosimeter photon response characteristics .............................................................12

6-2 Results of neutron spectrum measurements made about 24 in. in front of 

93%–96% HEU cylinders..........................................................................................................18

6-3 Historical distribution of deep dose equivalent..........................................................................20

6-4 Historical distribution of shallow dose equivalent......................................................................21
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 5 of 48

ACRONYMS AND ABBREVIATIONS

cm centimeter

cpm counts per minute

CTW construction trade worker

d day

DCF dose conversion factor

DOE U. S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

HEU highly enriched uranium

Hp(d) personal dose equivalent at tissue depth d (d = 10 mm or 0.07 mm)

hr hour

IARC International Agency for Research on Cancer

ICRP International Commission on Radiological Protection

ICRU International Commission on Radiation Units and Measurements

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1 thousand electron-volts

LOD limit of detection

MDL minimum detection level

MED Manhattan Engineer District (a DOE predecessor agency)

MeV megaelectron-volt, 1 million electron-volts

mg milligram

min minute

mm millimeter

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-OCAS Claims Tracking System

NTA nuclear track emulsion, type A (film)

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

OW open window

PGDP Paducah Gaseous Diffusion Plant

POC probability of causation

PORTS Portsmouth Gaseous Diffusion Plant

PPE personal protective equipment

QF quality factor
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RU recycled uranium

S shielded

SC&A Sanford Cohen & Associates

SEC Special Exposure Cohort

TBD technical basis document

TEPC tissue-equivalent proportional counter

TLD thermoluminescent dosimeter

TLND thermoluminescent neutron dosimeter

U.S.C. United States Code

yr year

§ section
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 7 of 48

6.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation 

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) restrict the “performance of duty” 

referred to in 42 U S. C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally-derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

in dose reconstruction. NIOSH, however, does not consider the following exposures to be 

occupationally derived (NIOSH 2010):

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 8 of 48

6.1.1 Purpose

The purpose of this technical basis document (TBD) is to provide technical data and other key 

information which will serve as the technical basis for evaluating external occupational dose for 

EEOICPA claimants who were employed at the Paducah Gaseous Diffusion Plant (PGDP).

6.1.2 Scope

PGDP workers, especially those employed during the peak production decades (1950s, 1960s, and 

1970s), have been exposed to radiation types and energies associated with enrichment of natural and 

recycled uranium (RU). PGDP used facility and individual worker monitoring methods to measure and 

control radiation exposure to workers (PGDP 1976). Before about July 1960, personnel dosimeters 

were not assigned to all workers (PACE and University of Utah 2000). Records of radiation dose to 

individuals who wore dosimeters are available beginning in 1953. Doses from these dosimeters were 

recorded at the time of measurement, routinely reviewed by PGDP operations and radiation safety 

personnel for compliance with radiation control limits, and routinely made available to individual 

workers. External Dose Reconstruction Implementation Guideline (NIOSH 2007) indicates that these 

represent the highest quality records for assessment and reconstruction of doses.

Initial radiation dosimetry practices were based on experience gained during several decades of 

radium and X-ray medical diagnostic and therapy applications. In general, these practices were well 

advanced at the start of the Manhattan Engineer District (MED) program to develop nuclear weapons, 

which began on August 13, 1942.

This TBD provides supporting technical data in the evaluation of occupational external dose for 

workers at PGDP. PGDP is one of the original sites that was designated by Congress as part of the 

Special Exposure Cohort (SEC) under EEOICPA [42 U.S.C. § 7384l(14)]. This designation is as 

follows:

(A) The employee was so employed for a number of work days aggregating at least 

250 work days before February 1, 1992, at a gaseous diffusion plant located in 

Paducah, Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment—

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges.

Dose reconstruction guidance in this TBD is presented to provide a technical basis for dose 

reconstructions for nonpresumptive cancers that are not covered in the SEC class through 

January 31, 1992. Dose reconstructions for individuals employed at PGDP before February 1, 1992, 

but who do not qualify for inclusion in the SEC, can be performed using this guidance as appropriate.

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 6.9.

6.2 BASIS OF COMPARISON

Since the start of the MED on August 13, 1942, various radiation dose concepts and quantities have 

been used to measure and record occupational dose. The basis of comparison for reconstruction of 

dose is the personal dose equivalent, Hp(d), where d identifies the depth (in millimeters) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of significance to 
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 9 of 48

skin dose, d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of significance to wholebody dose, d = 10 mm and is noted as Hp(10). Both Hp(0.07) and Hp(10) are the radiation quantities 

the International Commission on Radiation Units and Measurements (ICRU) has recommended for 

use as operational quantities for radiological protection (ICRU 1993). In addition, Hp(0.07) and 

Hp(10) are the radiation quantities the DOE Laboratory Accreditation Program (DOELAP) has used to 

accredit the Department’s personnel dosimetry systems since the 1980s (DOE 1986). The 

International Agency for Research on Cancer (IARC) Three-Country Combined Study (Fix et al. 1997) 

and the IARC Collaborative Study (Thierry-Chef et al. 2002) selected Hp(10) as the quantity to assess 

error in historical recorded whole-body dose for workers in IARC nuclear worker epidemiologic 

studies. This TBD uses Hp(10) and Hp(0.07) as deep dose and shallow dose, respectively.

6.3 DOSE RECONSTRUCTION PARAMETERS

Examinations of beta, photon (X- and gamma rays), and neutron energies and geometries of 

exposure, and the characteristics of PGDP dosimeter responses, are crucial for assessment of the 

original recorded doses. Bias and uncertainty for current dosimetry systems are typically well 

documented (Martin Marietta 1994). The performance of current dosimeters can often be compared 

to the performance of dosimetry systems in the same, or highly similar, facilities or workplaces. In

addition, current performance testing techniques can be applied to earlier dosimetry systems to 

achieve a consistent evaluation of all dosimetry systems. Dosimeter response characteristics for 

radiation types and energies in the workplace are crucial to the overall analysis of error in recorded 

dose.

Overall, accuracy and precision of the original recorded individual worker doses and their 

comparability to be considered in using NIOSH (2007) guidelines depend on the following factors (Fix 

et al. 1997):

• Administrative practices adopted by facilities to calculate and record personnel dose based 

on technical, administrative, and statutory compliance considerations

• Dosimetry technology, including physical capabilities of the dosimetry system, such as the 

response to different types and energies of radiation, in particular in mixed radiation fields

• Calibration of the respective monitoring systems and similarity of methods of calibration to 

sources of exposure in the workplace

• Workplace radiation fields that could include mixed types of radiation, variations in exposure 

geometries, and environmental conditions

The accuracy of PGDP worker doses has been the subject of DOE investigations (PACE and 

University of Indiana 2000). An evaluation of the original recorded doses as available, combined with 

detailed examinations of workplace radiation fields, is the recommended option to provide the best 

estimate of Hp(0.07) for the shallow dose and Hp(10) for the deep dose for individual workers.

6.3.1 Administrative Practices

The PGDP radiation monitoring program used portable instruments, contamination surveys, zone 

controls, and personnel dosimeters to measure exposure in the workplace (Harris 1957; PGDP 

1957a,b, 1964, 1976; UCND 1980). The program improved as better technology and more 

information became available. Results from personnel dosimeters were used to measure and record 

doses from external radiation exposure to PGDP workers. These dosimeters included one or more of 

the following:
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• Personnel whole-body beta/photon dosimeters

• Pocket ionization chamber dosimeters

• Personnel neutron dosimeters

For low-energy beta radiation, the dosimeters were probably incapable of furnishing accurate doses in 

terms of Hp(0.07). This TBD analysis does not include extremity doses, which were generally not 

assessed (PACE and University of Utah 2000).

In 1953, PGDP began using dosimeter and processing technical support provided by the Oak Ridge 

National Laboratory (ORNL) (Baker ca. 1995). There is evidence that PGDP might have processed 

its own dosimeters for a period; a review of the limited documentation available indicated that 

practices were similar to those used at ORNL and other major sites at that time (PGDP 1957a). 

Table 6-1 summarizes PGDP personnel beta/photon and neutron dosimeter characteristics [dosimeter 

type, exchange, minimum detection level (MDL), and potential missed annual dose]. ORNL, which 

was then the Clinton Laboratory, had based its dosimetry methods on the personnel beta/photon 

dosimeter design developed at the Metallurgical Laboratory at the University of Chicago (Pardue, 

Goldstein, and Wollan 1944). ORNL has provided PGDP with dosimeters from early in the operations 

period through the present. 

The precise detection levels listed in Table 6-1 are difficult to estimate, particularly for older systems. 

Current PGDP commercial thermoluminescent dosimeter (TLD) system MDLs are identified in ORNL 

documentation (Martin Marietta 1994) based on a DOELAP-accredited laboratory testing protocol 

(DOE 1986). During earlier years, MDLs were subject to additional uncertainty because factors 

involving radiation field and film type, as well as processing, developing, and reading systems, cannot 

now be tested (Thornton, Davis, and Gupton 1961). The estimates of film dosimeter MDLs in 

Table 6-1 were based on information from NIOSH (1993), NRC (1989), Wilson et al. (1990), and site 

personnel. Examination of older records, when available, indicated that the Hp(0.07) MDL values 

were about 3 times those for Hp(10) for film. The current TLD MDLs were obtained from ORNL 

(Martin Marietta 1994). The film badge was replaced by the TLD in 1980 (PGDP 1980). Parameters 

of the PGDP administrative practices significant to dose reconstruction involve policies to:

• Assign dosimeters to workers

• Exchange dosimeters

• Record notional dose (i.e., some identified value for lower dosed workers, often based on a 

small fraction of the regulatory limit) 

• Estimate dose for missing or damaged dosimeters

• Replace destroyed or missing records

• Evaluate and record dose for incidents

• Obtain and record occupational dose to workers for other employer exposure

PGDP policies appear to have been in place for all these parameters. From startup until July 1960, 

PGDP issued dosimeters to a limited number of individuals (PACE and University of Utah 2000). This 

population of monitored individuals represents those with the highest exposure potential. After July 

1960, PGDP routine practices required the assignment of dosimeters to all workers who entered a 

controlled radiation area (BJC 2000). Dosimeters were exchanged on a routine schedule (PGDP 

1957a, 1977; DOE 2000a). For workers in some areas the frequency was monthly, but for the 

general population it was quarterly. Employees on the monthly exchange cycle were primarily
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/2

4/2012 Page 11 of 48

Table 6-1. Dosimeter type, period of use, exchange frequency, MDL, and potential annual missed dose.

Dosimeter Period of use Monitored population

Exchange 

frequency

Laboratory 

MDL(rem)a

Maximum annual 

missed dose 

equivalent (rem)b

Hp(10) beta/photon dosimeters

Two-element film 1953 through 7/1960 Selected workers based on activities 

performed

Weekly (n = 50) 0.04 1.0

Four-element film After 7/1960 through 

1980

Workers in Buildings C-340, C-400, 

and C-410

Monthly (n = 12) 0.04 0.24

After 7/1960 through 

1980

Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.04 0.08 

After 7/1960 through 

1980

Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.04 0.02

Harshaw two-chip TLD 1980 through 1988 Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.02 0.04

Harshaw two-chip TLD 1980 through 1988 Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.02 0.01

Harshaw four-chip TLD, 

8800 series

1989 through present Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4)

Annual (n = 1)

0.02

0.02

0.04

0.01

Hp(0.07) beta/photon dosimeters

Two-element film 1953 through 7/1960 Selected workers based on activities 

performed

Weekly (n = 50) 0.12 3.0

Four-element film After 7/1960 through 

1980

Workers in Buildings C-340, C-400, 

and C-410

Monthly (n =12) 0.12 0.72

After 7/1960 through 

1980

Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.12 0.24

After 7/1960 through 

1980

Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual (n = 1) 0.12 0.06

Harshaw two-chip TLD 1980 through 1988 Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4) 0.03 0.06

Harshaw two-chip TLD 1980 through 1988 Workers and visitors not likely to 

exceed 0.1 of applicable guidelines

Annual ( n = 1) 0.03 0.015

Harshaw four-chip TLD, 

8800 series

1989 through present Workers and visitors with potential to 

exceed 0.1 of applicable guidelines

Quarterly (n = 4)

Annual ( n = 1)

0.02

0.02

0.04

0.01

Neutron dosimetersc

Harshaw TLND 1998 through present Selected workers based on activities 

performed

Quarterly (n = 4) 0.015 0.03

a. Estimated film dosimeter detection levels based on NIOSH (1993), NRC (1989), and Wilson et al. (1990). TLD detection levels from Martin Marietta (1994) and 

personal communication with site personnel.

b. Maximum annual missed dose (NIOSH 2007).

c. The potential annual missed dose based on laboratory irradiations is not applicable to workplace missed neutron dose.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 12 of 48

involved in chemical processing, maintenance of chemical processing facilities, and uranium metal 

production (DOE 2000a). All dosimeters were processed, and measured results were recorded and 

used to estimate dose.

Current administrative practices are generally available (Martin Marietta 1994), as is detailed 

information for each worker in the PGDP exposure history documentation. Summary documents 

provide information on historical practices at PGDP (PACE and University of Utah 2000; BJC 2000; 

PGDP 1957a, 1980; Baker ca. 1995).

6.3.2 Dosimetry Technology

PGDP dosimetry methods evolved with the development of improved technology and better 

understanding of complex radiation fields. The adequacy of dosimetry methods to measure radiation 

dose accurately is determined from radiation type, energy, exposure geometry, and other factors 

described in this section. The dosimeter exchange frequency gradually lengthened, corresponding in 

general to the period of regulatory dose controls.

6.3.2.1 Beta/Photon Dosimeters

PGDP has historically used personnel dosimeter services from ORNL. In 1945, ORNL implemented 

the beta/gamma film dosimeter design, which was developed originally at the Metallurgical Laboratory 

at the University of Chicago (Pardue, Goldstein, and Wollan 1944). ORNL followed a research and 

development process that led to gradual upgrades in dosimetry capabilities for complex radiation 

fields (Thornton, Davis, and Gupton 1961). Other DOE sites followed this evolution in dosimetry 

capabilities, which led to site-specific multielement film and thermoluminescent dosimetry systems.

Figure 6-1 shows the energy response characteristics of the PGDP beta/gamma dosimeters based on 

the essentially identical two-element film dosimeter designed at the University of Chicago and used at 

the Hanford Site (as well as ORNL, Los Alamos National Laboratory, and probably other MED sites). 

Figure 6-1. Estimated dosimeter photon response characteristics.

In addition, Figure 6-1 shows the Hp(10) response. Further, the figure shows the energy response of 

Hanford multielement film and TLDs (Wilson et al. 1990). The curve labeled “Two-Element Film 

Shield” represents ORNL dosimeters from 1945 through 1978. ORNL used a multielement film 

dosimeter after 1953 (Thornton, Davis, and Gupton 1961), but processed photon response as it did 

for the two-element dosimeter and used the same shielding as that used in the two-element 

dosimeter. The figure shows that the two-element dosimeter over-responded in relation to Hp(10)
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 13 of 48

from 0.05 to 0.3 MeV, followed Hp(10) for higher energies, and under-responded for lower energies. 

It also shows that TLDs are capable of following Hp(10) over the energy range of interest. The 

majority of PGDP worker photon dose comes from handling uranium of low enrichment. The photon 

energy spectrum is almost entirely in the range from 30 to 250 keV (Schleien et al. 1998).

The nonpenetrating response of the two-element dosimeter was calculated as the difference between 

the unshielded and shielded portions of the film based on a uranium calibration. The two-element 

dosimeter workplace nonpenetrating (i.e., beta or shallow) dose response based on the uranium 

calibration should adequately represent Hp(0.07) or at least be favorable to claimants because of the 

significant over-response of the unshielded portion of the film to any lower energy photons that could 

have been present (Wilson et al. 1990). The multielement film dosimeters and TLDs, which were also 

calibrated to uranium slabs, had the ability to correct more accurately for mixed photon and beta 

radiation (Wilson et al. 1990).

6.3.2.2 Neutron Dosimeters

Dosimeters used at PGDP historically had a neutron-sensitive element that was processed on 

request. After 1989, this capability has been provided with a TLD that contained a 6

LiF chip, which is 

very responsive to low-energy neutrons. There is no indication of recorded neutron doses for PGDP 

workers wearing either of these dosimeters [1]. The use of commercial Harshaw thermoluminescent 

neutron dosimeters (TLNDs) to assess neutron dose routinely (along with deep and shallow dose) 

began in 1998. ORNL has provided the dosimeters and associated services. The albedo dosimeter 

has been worn with a belt to minimize distance from the worker’s body, which optimizes the albedo 

effect for which the dosimeter is calibrated [2].

The quality factors (QFs) used historically for neutrons have changed significantly. In current 

regulations, QFs that are used to convert radiation dose (millirad) to dose equivalent (millirem) are 

based on International Commission on Radiological Protection (ICRP) Publication 38 (ICRP 1983). 

The most current QFs from ICRP Publication 90 (1991) are about 2 times higher than the ICRP 

(1983) values. Because a QF of 10 was used for the referenced radiation measurements, the PGDP 

personnel dosimetry, an adjustment to ICRP (1991) of at most a factor of 2 times higher, would be 

necessary [3].

Average neutron energy is less than about 1 MeV, 510 keV for 2% 235U, 770 keV for 5% 235U, and 

860 keV for 97% 235U (Cardarelli 1997, p. 9). QF equals 10 for ICRP (1983), or about 20 for the ICRP 

(1991) revision. The average neutrons from depleted and natural uranium cylinders ranged from 210 

to 360 keV (Cardarelli 1997, p. 9). Unmoderated and deuterium (water) 252Cf neutrons created were 

between 1,306 and 1,403 keV. This means the dose as monitored at PGDP since 1998 was 

overestimated and, therefore, is favorable to the claimant.

6.3.3 Calibration

Potential error in recorded dose is dependent on dosimetry technology response characteristics to 

each radiation type, energy, and geometry; the methodology used to calibrate the dosimetry system; 

and the extent of similarity between the radiation fields used for calibration and that present in the 

workplace. The potential error is much greater for dosimeters with significant variations in response, 

such as film dosimeters for low-energy photon radiation and the nuclear track emulsion and TLNDs

for neutron radiation [4].

6.3.3.1 Beta/Photon Dosimeters

The beta/photon film dosimeters at PGDP were calibrated to 226Ra until 1980, when the calibration 

source changed to 137Cs (ORAUT 2007b). The calibration to both 226Ra and 137Cs was free in air (no 
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 14 of 48

phantom) until the DOELAP procedures adopted in 1986 required phantoms (DOE 1986). Hp(10) is 

defined with a phantom, in particular the ICRU slab phantom, which is a conservative practical 

definition of anterior-posterior whole-body dose to the standard ICRU spherical phantom (ICRU 1993).

Introduction of on-phantom calibration of film dosimeters and replacement of 226Ra by 137Cs as the 

calibration source changed the relationship between recorded dose and Hp(10). In addition to 

registration of the additional backscattered radiation, the generally lower energy photon spectrum 

from 226Ra in comparison with that from 137Cs (662 keV) gave a greater optical density for the same 

dose during calibration (Figure 6-1). In contrast, the effect of backscatter is to overestimate dose, and 

calibration with 226Ra tends to underestimate the dose in relation to calibration with 137Cs. 

In the 1980s, studies at a number of laboratories assessed changes from the on-phantom calibration 

mandated by the DOELAP testing criteria (Fix et al. 1982; Wilson 1987; Wilson et al. 1990; Taylor et 

al. 1995). While not exactly the same at all sites, most film dosimeters, like those at PGDP, had 

common features due to their evolution from the original work of Pardue, Goldstein, and Wollan 

(1944). The early badges were calibrated to exposure in free air. Laboratory tests at the Hanford Site 

showed 8% and 4% increases in dosimeter response for on-phantom exposures using 226Ra and 

137Cs, respectively (Fix et al. 1982). With free-air calibration, the exposure to the wearer tends to be 

overestimated by this amount, which is assumed to be similar for Paducah. Tests at the Savannah 

River Site, on the other hand, indicated that film badge doses underestimated Hp(10) by 11.9% before 

1986 and by 3.9% in 1986 (Taylor et al. 1995). Lacking site-specific data for PGDP, this TBD 

recommends the use of exposure-to-organ dose conversion factors (DCFs) in NIOSH (2007, 

Appendix B) for dose reconstruction at PGDP with no numerical adjustment to the recorded doses; 

this procedure should be favorable to claimants (Fix et al. 1982). It allows for an overestimate of 

exposure, as assessed in the Hanford studies, that should be sufficient to offset effects due to the 

calibration source if they are in the opposite direction.

For a number of years, ORNL used uranium beta as well as 226Ra gamma calibration curves to 

interpret film densities (Thornton, Davis, and Gupton 1961). The ratio of beta-to-gamma responses 

was tested in several ways. Films wrapped in a 7-mg/cm2 absorber were placed in contact with a slab 

of natural uranium. The densities per rad were nearly the same as those produced from 226Ra gamma 

rays measured behind a cadmium filter. In addition, stacks of film were exposed on a uranium 

surface, and the densities at various depths were used to extrapolate to the value for a depth of 

7 mg/cm2

. This value was nearly equal to that produced by the same dose from 226Ra photons behind 

the cadmium filter. Therefore, for beta radiation from natural uranium, the density produced per rad in 

film was equal to the density produced per rad behind the cadmium filter by 226Ra gamma rays. 

Analysts concluded that, for routine personnel dosimetry, film was equally sensitive for beta and 

gamma radiations. Because the film badge had a minimum absorber thickness of 80 mg/cm2

between the film and the source, the effective beta energy is needed to interpret the film density in 

terms of Hp(0.07). The radiation was routinely treated as 1.7-MeV beta particles from uranium, which 

are about 40% absorbed in 80 mg/cm2 (Thornton, Davis, and Gupton 1961). Thus, the determination 

of beta dose was specific to uranium.

6.3.3.2 Neutron Dosimeters

Calibration of neutron dosimeters for use at PGDP was appropriate for the work locations in which 

those dosimeters were worn (Martin Marietta 1994). Dosimeter response was characterized in a 

manner that would represent the workplace (Martin Marietta 1994). Reference dosimetry for these 

measurements was evaluated with tissue-equivalent proportional counters (TEPCs). TEPCs provide 

an absolute measure of absorbed dose in a tissue-like material and, with an appropriate algorithm, an 

estimate of the neutron QF (Scherpelz and Murphy 1995). The basis for the calibration factor was 

developed using data obtained at the Y-12 Plant in a room used to store an array of small canisters of 
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 15 of 48

UF4. Measurements were made with Bonner spheres at the same location. The average QF was 11, 

and the average energy range was 0.6 to 1.4 MeV (PNL 1990).

In 1989, field measurements for neutron flux were made by PNL representatives at the end row of the 

cylinder yard at the K-25 plant. The measurements were completed with a TEPC and a phantom with 

TLDs approximately 4 ft from the outside of a cylinder; the phantom was near the center of the 

cylinder’s length. The results were evaluated qualitatively because the dose rate was low and an 

appropriate power supply was not available. The calibration factors were similar to those in Y-12 

Building 9212 in the UF4 storage area container array and confirmed the appropriateness of these 

values (PNL 1990). These calibration factors apply to the PGDP TLNDs (Martin Marietta 1994).

6.3.4 Workplace Radiation Fields

6.3.4.1 Beta/Photon Fields

PGDP operations are characterized by the relatively low-level external beta and photon radiation 

fields associated with uranium in feed materials, products, wastes, and contaminated equipment and 

systems. Processed RU was present with natural, depleted, and enriched (up to 2% 235U by weight) 

abundances. (Section 6.3.4.3 describes potential sources for neutron exposure.)

Table 6-2 summarizes the major sources of external radiation throughout PGDP operations (PACE 

and University of Utah 2000). The photon energy range of principal interest is 30 to 250 keV. 

Handling uranium material of these types did not, in general, produce areas with significantly elevated 

photon radiation. 

Table 6-2. Major radiation sources.

Nuclide Source Half-life

Energies (MeV) and abundances 

of major radiations

Alpha Beta (max) Gamma

U-238 Primary U isotope 4.51E9 yr 4.15 (21%)

4.20 (79%)

U-235 Primary U isotope 7.1E8 yr 4.21 (6%) 0.144 (11%)

4.37 (17%) 0.163 (5%)

4.40 (55%) 0.186 (57%)

4.60 (5%) 0.205 (5%)

U-234 Primary U isotope 2.47E5 yr 4.72 (28%) 0.053(0.12%)

4.77 (72%)

Th-234 Decay product 24.1 d 0.013 (9.8%)

0.103 (21%) 0.063 (3.5%)

0.193 (79%) 0.092 (3%)

0.093 (4%)

Pa-234m Decay product 1.17 min 2.29 (98%) 0.765 (0.3%)

1.001 (0.60%)

Th-231 Decay product 25.5 hr 0.206 (13%)

0.287 (12%) 0.026 (2%)

0.288 (37%) 0.084 (10%)

0.305 (35%)

Tc-99 Impurities from RU 2.12E5 yr 0.294 (100%) None

The major facilities and associated activities at PGDP are (BJC 2000):

• C-331, C-333, C-335, and C-337 – Gaseous Diffusion Process Buildings

• C-410/420 – UF6 Feed Plant

• C-310 – Purge and Product Withdrawal Building
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 16 of 48

• C-315 – Surge and Tails Withdrawal Building

• C-340 – Metals Plant

• C-400 – Decontamination and Cleaning Building

• C-720 – Maintenance Building

The buildings with the greatest potential for elevated direct radiation levels were C-340, C-410, C-420, 

and the cascade buildings (PACE and University of Utah 2000). From 1952 to approximately 1980, 

the major sites of potential exposure to radioactive material were buildings involved in the conversion 

of UO3 powder to enriched UF6 in solid or gaseous form, UF4 and uranium metals recovery operations, 

and the decontamination building. Feed and enrichment operations were in Buildings C-410, C-420, 

C-331, C-333, C-335, C-337, C-310, and C-315, while UF4 and uranium recovery were in Building 

C-340 (PGDP 1957b). The decontamination operation was in Building C-400. The oxide conversion 

building, C-420, was where UO3 powder (clean or recycled) was received and converted to UF4. 

From Building C-420, material went to Building C-410, the feed plant, for conversion to UF6. Last, UF6

was processed through the cascade buildings (C-331, C-333, C-335, and C-337). Enriched UF6 was 

withdrawn in Building C-310, the product withdrawal building, while depleted UF6 was removed in 

Building C-315, the tails withdrawal building. Radiation surveys were performed near the UF6

cylinders to evaluate the potential for exposure to personnel working adjacent to the shipping 

containers and area exposure rates in the cylinder yards (McDougal 1980; Frazee 1982; Mason 

1986). Table 6-3 lists the principal buildings, sources for external dose, and periods of operation.

Table 6-3. Buildings and periods of operation.

Site facilities Source for external dose

Operation

Begin End

C-310 Purge and Product 

Withdrawal

UF6 process equipment and cylinders 1953 Ongoing

C-315 Surge and Tails Withdrawal UF6 process equipment and cylinders 1953 Ongoing

C-331, C-333 Gaseous Diffusion 

Process Buildings

UF6 process equipment and cylinders 1952 1964

1969 1970

1972 1976

C-335, C-337 Gaseous Diffusion 

Process Buildings

UF6 process equipment and cylinders 1954 1964

1969 1970

1972 1976

C-340 Reduction and Metals 

Facility

Process equipment, contaminated floors 1957 1962

1967 1977

C-400 Decontamination and 

Cleaning Buildings

UF6 process equipment and cylinders 1952 1990

C-410 UF6 Feed Plant and C-420 

Oxide Conversion Plant

Process equipment, contaminated floors 1953 1964

1968 1977

C-415 Feed Plant Storage Building Radioactive source storage area 1953 1977

C-745 A-V Cylinder Yards UF6 cylinders 1953 (estimated) Ongoing

PGDP also processed RU. The feed material contained trace amounts of radioactive impurities not 

present in natural uranium feed material. Because these impurities were present in such minute 

concentrations, their radiological impact was usually negligible (PACE and University of Utah 2000). 

However, some routine chemical processes would concentrate them (PACE and University of Utah 

2000). From an external dose standpoint, the most significant impurity in RU is the pure beta emitter, 

99Tc, which tends to deposit in enrichment equipment and pocket in the higher sections of the 

diffusion cascade (DOE 2000b). Technetium-99 was also concentrated for recovery and removal. 

The relatively low-energy beta particles (maximum 294 keV) from 99Tc pose minimal external 

exposure potential because of their limited range. Neither film nor TLD efficiently detects them, 

particularly in the presence of uranium. Clothing and gloves provide adequate shielding. Skin 

contamination is the only credible scenario in which significant shallow dose could occur from 99Tc. 
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 17 of 48

6.3.4.2 Workplace Beta/Photon Dosimeter Response

Essentially all PGDP radiological work areas involved photon and beta radiation characteristic of 

operations involving uranium at low enrichments. As discussed in Section 6.3.3.1, the recorded 

responses of the PGDP beta/photon film dosimeters are favorable to claimants and need no 

adjustment.

6.3.4.3 Neutron Fields

While neutrons occur in some areas at PGDP, measured levels are low. There are no identified 

locations where measurable neutron dose was encountered (Martin Marietta 1994). Studies have 

evaluated neutron fields at gaseous diffusion plants (Scherpelz and Murphy 1995; Cardarelli 1997); 

these studies confirm Martin Marietta (1994). Cylinder yards, feed and withdrawal areas, and 

locations where uranium forms deposits in the cascade have been investigated (Cardarelli 1997). 

These studies identified the storage cylinders, which contained either depleted UF6 (tails) or enriched 

UF6 (product), as areas where neutron fields could represent an exposure hazard. Estimates of dose 

equivalent rates range from 0.007 to 0.34 mrem/hr; associated QFs range from 7 to 10. Radiation 

measurements indicated that the neutron flux increased as a function of uranium enrichment; neutron 

flux increased from 0.2 mrem/hr for cylinders with as much as 5% enrichment to 4 mrem/hr on contact 

with 97% enrichment (DOE 2000b). A representative average value is 0.2 mrem/hr based on a QF of 

about 10 (Scherpelz and Murphy 1995; Cardarelli 1997). Estimates of average neutron energies 

ranged from 0.25 to 0.56 MeV (Scherpelz and Murphy 1995). Neutron monitoring of individuals was 

performed during a UF6 cylinder-painting project (BJC 1999). Results of this project indicated a 

neutron-to-photon dose equivalent ratio of approximately 1 to 5, based on a QF of 10 [5].

Cylinders of highly enriched (93% to 96%) uranium (HEU) were measured with a TEPC mounted on a 

phantom about 24 in. from the cylinders (Soldat and Tanner 1992). The dose equivalent from the 

cylinders was about 0.8 mrem/hr with a total dose equivalent of 14 mrem. The multisphere 

measurement at the same location as the phantom resulted in an average neutron energy of 

0.53 MeV and a dose equivalent rate of 0.5 mrem/hr.

The solid lines in Figure 6-2 show the calculated energy spectrum from the multisphere detectors 

(Bonner spheres). Table 6-4 lists dose fractions for the neutron energy groups (indicated by the 

dashed lines in Figure 6-2). The dose fractions for the lower (less-than-10-keV) and intermediate (10-

to 100-keV) energy neutron groups were about 47% of the total dose from the measurements 

(ORAUT 2007b).

Exposure to low enriched UF6 (less than 5%) will result in a lower neutron flux than the neutron field 

expected from highly enriched UF6 (greater than 97%) as surveyed at the Portsmouth Gaseous 

Diffusion Plant (PORTS) by Soldat and Tanner (1992). The dose fractions listed in Table 6-4 are 

favorable to claimants (Soldat and Tanner 1992).

The neutron study performed in 1990 at X-10 and Y-12 (Soldat et. al. 1990) was the only definitive 

study of neutron energy spectra documented over the history of PGDP. It is assumed that the energy 

spectra are valid for the earlier years, given the presence of enriched uranium.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 18 of 48

Figure 6-2. Results of neutron spectrum measurements 

made about 24 in. in front of 93%–96% HEU cylinders

(Soldat and Tanner 1992).

Table 6-4. Dose fractions for PORTS HEU storage vault 

in Building 345.

Neutron energy group Near unshielded Cf-252 source

< 10 keV 0.300

10-100 keV 0.172

0.1-2 MeV 0.447

2-14 MeV 0.081

Dose fractions that are favorable to claimants

< 10 keV 0.300

0.1-2 MeV 0.610

2-20 Mev 0.081

6.3.4.4 Workplace Neutron Dosimeter Response

Quantitative monitoring for neutron dose began at PGDP in 1998. TLNDs were used in conjunction 

with appropriate work field calibration factors. Before 1998, the beta/photon badge assembly 

contained a neutron-sensitive element [nuclear track emulsion, type A film (NTA); Eastman Kodak 

Type 2 film]. This element was processed only when requested. (NTA film had an energy threshold 

of about 0.5 MeV.) A review of data does not indicate the assignment of neutron dose before 1998.

6.4 ADJUSTMENTS TO RECORDED DOSE

6.4.1 Photon Dose

Recorded doses varied in reporting units depending on regulatory requirements and dose definitions 

(national and international). The reporting unit used by DOE is the millirem, a unit of dose equivalent. 

The international unit of dose equivalent is the millisievert, which is equivalent to 100 mrem. Since 

1986, deep dose equivalents at PGDP have been based on DOELAP calibration to Hp(10) and 

require no adjustment. Before 1986, TLDs were calibrated in air to 137Cs, which is nearly equivalent to 

an Hp(10) on-phantom 137Cs calibration. No adjustment to the measured TLD penetrating photon 

dose is necessary. As discussed in Section 6.3.3.1, the earlier film badge deep doses are favorable 

to claimants and require no numerical adjustment.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 19 of 48

6.4.2 Nonpenetrating Dose

The early film dosimeters were calibrated to uranium for nonpenetrating radiation. No numerical 

adjustment of recorded shallow doses is recommended. Incident reports are a possible source that 

dose reconstructors can consult for investigations of nonroutine beta exposures and dose 

assessment.

6.4.3 Neutron Dose

Measured neutron energies at PGDP are between 0.10 and 2.0 MeV, for which the ICRP Publication 

60 radiation weighting factor is 20 (ICRP 1991). Therefore, dose reconstructors should multiply the 

reported neutron dose equivalent by the appropriate ICRP (1991) correction factor to be used for 

reconstruction (NIOSH 2007). Apply this factor to measured, missed, and unmonitored neutron 

doses.

6.5 MISSED DOSE

Missed deep and shallow doses have been examined for three groups of PGDP workers as follows:

1. A zero dose was recorded but the worker was not monitored (most workers from 1953 to July 

1960).

2. A zero dose was recorded for the dosimeter system for any response less than the MDL.

3. There was no recorded dose because workers were not monitored or the dosimetry record is 

not available. 

Neutron dose rates at PGDP were low (Martin Marietta 1994). Neutron dosimeters were not routinely 

assigned and doses were not recorded until about 1998. Neutron doses reported before 1998 were 

based on a conservative calibration associated with a neutron-sensitive element in the beta/gamma 

dosimeter. Application of a neutron-to-gamma dose equivalent ratio of 1 to 5 appears to be a 

satisfactory option that is favorable to claimants because the photon dose is reliably measured. This 

ratio can be applied to selected work activities [6].

6.5.1 Estimating Missed and Unmonitored Photon Deep Dose

Watson et al. (1994) examined methods to be considered when there is no recorded dose for a period 

during a working career. In general, estimates of unmonitored dose can be made by using dose 

results for coworkers or the recorded dose before and after the period when they were not monitored. 

However, these situations require careful examination. The dose reconstructor should consider all 

reasonable methods and assign the most appropriate dose based on employee job description and 

work locations. NIOSH (2007) cites several different models.

For Group 2, the missed dose for dosimeter results that are less than the MDL is particularly important 

for earlier years, when MDLs were higher and dosimeter exchange was more frequent. NIOSH 

(2007) describes an acceptable estimate that is favorable to claimants of the maximum potential 

missed dose as one-half the MDL multiplied by the number of zero dose results (the MDL/2 method). 

The right-hand column in Table 6-1 lists estimates of the annual missed dose for Group 2 at PGDP.

If it is definite that the employee was not a radiation worker, the unmonitored deep dose for that 

period can be assigned as the onsite ambient dose.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 20 of 48

Otherwise, dose reconstructors should treat an individual in Group 1 or 3 as a radiation worker, then 

approach the unmonitored deep dose in two ways. First, consider the same assignment of missed 

dose as that for Group 2, from the right-hand column of Table 6-1. However, for 1953 through July

1960, with the frequent (weekly) dosimeter exchange and relatively large MDL, the resulting implied 

annual missed dose of 1 rem is probably unrealistically large for many unmonitored persons in 

Groups 1 and 3. Figure 6-3 shows the distribution of individual annual deep dose equivalent for 

monitored workers from 1953 to 1974 (Baker ca. 1995). Few of these individuals received as much 

as 1 rem in a year.

Figure 6-3. Historical distribution of deep dose equivalent (Baker ca. 1995).

An alternative approach for Group 1 or 3 is to base the unmonitored dose estimate on exposure data 

compiled for monitored PGDP workers. ORAUT-OTIB-0020, Technical Information Bulletin: Use of 

Coworker Dosimetry Data for External Dose Assignment (ORAUT 2011a), provides general 

instructions to evaluate the measured and missed doses for monitored PGDP workers to arrive at a 

dose to be assigned to unmonitored workers that is favorable to claimants. Attachment B contains the 

details of the evaluation of PGDP coworker dose to be assigned to unmonitored workers. These 

measured doses do include an analysis of the missed dose, which is particularly significant for the 

earlier years with higher Limits of Detection (LODs) and frequent dosimeter exchanges. 

Attachment B, Table B-2 provides the 50th- and 95th- percentile coworker doses.

6.5.1.1 Construction Trade Workers

Construction Trade Worker (CTW) measured doses are increased to account for uncertainty for 

reasons described in ORAUT-OTIB-0052, Technical Information Bulletin: Parameters to Consider 

When Processing Claims for Construction Trade Workers (ORAUT 2011b). For extended 

employment periods without a measured dose, consideration about whether to assign an unmonitored 

dose using the coworker doses in Attachment B is necessary. In this case, the measured coworker 

penetrating annual dose has been multiplied by a factor of 1.4 (ORAUT 2011b) and the missed dose 

determined using NIOSH (2007) guidance. Attachment B, Table B-3 lists the 50th- and 

95th-percentile CTW doses to be assigned. 

6.5.2 Estimating Missed and Unmonitored Shallow Dose

The procedure for assessing missed and unmonitored shallow dose is similar to that for missed deep 

dose. 
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 21 of 48

For Group 2, the last column of Table 6-1 lists the missed annual shallow dose equivalent in keeping 

with the MDL/2 method of evaluation. Guidance on determining the reconstructed skin dose can be 

obtained from Attachment B for PDGP workers. Figure 6-4 shows the historical data for the 

distribution of shallow dose equivalent for monitored workers at PGDP (Baker ca. 1995). When 

compared with Figure 6-4, this assessment of annual missed shallow dose for Group 2 is favorable to 

claimants.

Figure 6-4. Historical distribution of shallow dose equivalent (Baker ca. 1995).

For nonradiological workers in Groups 1 and 3, the unmonitored shallow dose can be assigned as the 

environmental dose. Dose reconstructors should regard other individuals in these groups as radiation 

workers, and consider the same estimate as that used for Group 2. As an alternative, use 

Attachment B, Table B-2. Significant nonroutine beta doses, as from skin contamination events 

(particularly during 99Tc recovery and removal), could be addressed in specific incidence reports. 

Attachment A provides guidance.

6.5.3 Estimating Missed and Unmonitored Neutron Dose

Dose reconstructors should add a neutron component to the annual dose of individuals prior to 1998. 

Table 6-5 lists the criteria for assigning missed and unmonitored neutron dose. Radiological areas 

are, but not limited to, cascade facilities, feed and production withdrawal areas, oxide conversion 

facilities, feed manufacturing facilities, decontamination/cleaning facilities, cylinder yards, and neutron 

source storage areas. Beginning in 1998, employees with neutron exposure were monitored and 

these dosimetry records should be used for dose assignment. The neutron dose equivalent should be 

multiplied by a factor of 2 as previously discussed in Section 6.4.3.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 22 of 48

Table 6-5. Missed and unmonitored neutron dose assignment.

Description Neutron dose assignment

Unmonitored employee – all radiological areas Assign neutron dose based on coworker dose. Apply a 

neutron-to-photon ratio of 0.2 for dose equivalent (BJC 

1999).a

Monitored (photon and electron) employee Assign neutron dose if positive photon results are 

recorded. Apply a neutron-to-photon ratio of 0.2 for 

dose equivalent (BJC 1999).a

Monitored (photon and electron) employee having 

no dosimetry results greater than half the limit of 

detection

Assign neutron dose based on the missed dose. Apply 

a neutron-to-photon ratio of 0.2 for dose equivalent 

(BJC 1999).a

a. BJC (1999)

6.6 UNCERTAINTY

PGDP has historically used ORNL personnel dosimeter services. ORNL has assessed the standard 

error in the recorded film-badge dose as ±30% for photons of all energies (ORAUT 2007a). The 

standard error for beta dose is the same (or somewhat larger for unknown mixtures of beta/gamma 

dose). Thus, the film badge dose uncertainty is 1.3. The uncertainty in the TLD dose is 1.15 (ORAUT 

2007a), which is consistent with NIOSH (2007).

6.7 DOSE RECONSTRUCTION

As much as possible, dose reconstructors should base dose to individuals on dosimetry records. It is 

important to distinguish between the recorded nonpenetrating and penetrating doses and the actual 

Hp(0.07) and Hp(10). The following list summarizes appropriate information for dose reconstructors:

• Consider dosimetry records that provide nonzero beta-photon values for Hp(10) and Hp(0.07)

to be adequate. No numerical adjustment of the doses is required. Beta energies are greater 

than 15 keV and photon energies are in the range from 30 to 250 keV.

• Assign missed dose to workers for whom dosimetry records provide zero beta-photon values 

for Hp(10) and Hp(0.07) on the basis of MDL/2 times the number of zero results, as described 

in Sections 6.5.1 and 6.5.2 (NIOSH 2007).

• Individuals with no dose recorded might or might not have been radiological workers. If it is 

definite that the individual was not a radiation worker, the assigned missed dose is the 

environmental dose discussed in the Occupational Environmental Dose part of this Site Profile 

(ORAUT 2012b). Otherwise, estimate the missed dose as described in Section 6.5. No 

numerical adjustments to the missed dose are necessary.

• Multiply reported, missed, and unmonitored neutron dose equivalents by the appropriate ICRP 

(1991) correction factor.

• Base the assignment of missed and unmonitored neutron dose equivalent estimate on a 

neutron-to-photon ratio of 0.2 for dose equivalent (BJC 1999) prior to 1998. Beginning in 1998 

base the neutron assignment on the dosimetry records. Assign missed neutron dose using 

Table 6-1. Multiply the estimated neutron dose equivalent by 2 to adjust for ICRP (1991).

• Pay special attention to the possibility of skin contamination incidents for workers involved with 

99Tc recovery operations (Attachment A).

• See Section 6.6 for a discussion of uncertainty.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 23 of 48

6.8 ORGAN DOSE

NIOSH (2007) discusses the conversion of measured doses to organ dose equivalent, and 

Appendix B of that document contains the appropriate DCFs for each organ, radiation type, and 

energy range based on the type of monitoring performed. In some cases, simplifying assumptions are 

appropriate [7].

6.9 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data, 

quotations, and other information to documents available for review on the Oak Ridge Associated 

Universities Team servers.

[1] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

The reviewed records did not reveal recorded neutron doses for either dosimeter.

[2] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

Proximity of the albedo dosimeter is important for its response; standard practice ensured this.

[3] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

ICRP (1991) recommended weighting factor of 20 for neutron energies between 0.1 and 

2 MeV. Doses of record used a QF of 10; therefore, a factor of 2 correction is indicated.

[4] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

The importance of energy response to accurate measurement of dose equivalent is well 

known, and the response of the historical dosimeters is shown in Figure 6-1.

[5] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

The determination of a neutron-to-photon ratio for absorbed dose was based on a dose 

equivalent ratio that can be used to estimate neutron dose from photon measurements. 

[6] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

Empirical neutron and photon worker dose equivalent data provide a basis from which neutron 

dose equivalent can be inferred from better known photon dose. Interpretation of BJC (1999). 

[7] Turner, James E. Integrated Environmental Management. Consultant. 2003. 

Appendix B of NIOSH (2007) contains tables for numerous organs. Some professional 

judgment is needed to fit particular conditions.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 24 of 48

REFERENCES

Baker, R. C., ca. 1995, Occupational Radiation Exposure Experience, Paducah Gaseous Diffusion 

Plant, Union Carbide Corporation, Nuclear Division, Paducah, Kentucky. [SRDB Ref ID: 

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GLOSSARY

absorbed dose

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays. See dose.

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

albedo effect

In relation to health physics, dosimeter response caused by the moderating and 

backscattering of neutron radiation by a human chest or a phantom.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

attenuation

Process by which absorption and scattering reduces the number of particles or photons 

passing through a body of matter.

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most direct fission products are (negative) beta emitters. Exposure 

to large amounts of beta radiation from external sources can cause skin burns (erythema), and 

beta emitters can be harmful inside the body. Thin sheets of metal or plastic can stop beta 

particles.

Bonner sphere

See multi-sphere neutron spectrometer.

cascade

At PGDP, series of compressor, heat exchanger, control valve and motor, converter stages, 

and supporting piping arranged in stages, cells, and units that progressively increase the 

concentration of 235U in a uranium hexafluoride (UF6) feed. Enrichment occurs as UF6 passes 

through semiporous barriers in the converter stage. These barriers allow the lighter 235U 

molecules to pass through more easily, which results in a gas with a slightly higher percentage 

of 235U (enriched) on one side of the barrier and a slightly lower percentage (depleted) on the 

other side. The enriched UF6 gas flows toward the top of the cascade while the depleted UF6

gas travels toward the bottom of the cascade.
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curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

deep dose equivalent (Hd)

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter). See dose.

DOE Laboratory Accreditation Program (DOELAP)

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs.

dose

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. Various terms narrow the type of dose, 

and some are additive: 

• Absorbed dose is the amount of energy deposited in a substance by ionizing radiation.

• Deep dose is the dose at a 1-centimeter depth in tissue (1,000 milligrams per square 

centimeter).

• Effective dose is the sum of the equivalent doses in the principal tissues and organs of 

the body, each weighted by a tissue weighting factor that accounts for the probabilities 

of fatal and nonfatal cancers according to severity and the average length of life lost 

due to an induced cancer. It indicates the biological effect of the radiation exposure in 

that tissue. 

• Equivalent dose or dose equivalent is the absorbed dose in a tissue or organ multiplied 

by a weighting factor for the particular type of radiation. 

• Organ dose is the dose to a specific organ.

• Penetrating dose is that from higher energy photon (gamma and X-ray) radiation and 

neutron radiation that penetrates the outer layers of the skin. Nonpenetrating dose is 

that from beta and lower energy photon radiation.

• Personal dose equivalent is the dose equivalent in soft tissue below a specified point 

on the body at a specified depth.

• Shallow dose is the dose at a 0.07-centimeter depth in tissue (7 milligrams per square 

centimeter).

• Skin dose is the dose to the skin.

• Whole-body dose is the dose to the entire body excluding the contents of the 

gastrointestinal tract, urinary bladder, and gall bladder.
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dose conversion factor (DCF)

Multiplier for conversion of potential dose to the personal dose equivalent to the organ of 

interest (e.g., liver or colon). In relation to radiography, ratio of dose equivalent in tissue or 

organ to entrance kerma in air at the surface of the person being radiographed.

dose equivalent (H)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, neutron film dosimeter, 

pocket ionization chamber, and thermoluminescent dosimeter.

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results. 

enrichment

Isotopic separation process that increases the percentage of a radionuclide in a given amount 

of material above natural levels. For uranium, enrichment increases the amount of 235U in 

relation to 238U. Along with the enriched uranium, this process results in uranium depleted in 

235U. At PGDP this involves a process that occurs as UF6 passes through barriers in 

converters allowing isotopes of lower molecular weight to pass through.

external dose

Dose received from radiation emitted by sources outside the body.

film

Radiation-sensitive photographic film in a light-tight wrapping. See film dosimeter.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 
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gaseous diffusion plant

Facility where uranium hexafluoride (UF6) gas is filtered to enrich the 235U and separate it from 

238U. The process requires enormous amounts of electric power and results in an increase in 

235U enrichment from 1% to about 3%.

gray (Gy)

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 gray equals 1 joule per kilogram or 

100 rads.

highly enriched uranium (HEU)

Uranium enriched to at least 20% 235U for use as fissile material in nuclear weapons 

components and some reactor fuels. Also called high-enriched uranium.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation.

limit of detection (LOD)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level. See minimum detectable level.

minimum detectable activity

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

minimum detectable level (MDL)

See minimum detectable activity.

multi-sphere neutron spectrometer

Spectrometer that consists of a series of neutron-moderating spheres of tissue-equivalent 

material with a neutron detector in the middle of the respective spheres. Algorithms are used 

to calculate the neutron spectra.

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.
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nonpenetrating dose

Dose from beta and lower energy photon (X-ray and gamma) radiation that does not penetrate 

the skin. It is often determined from the open window dose minus the shielded window dose. 

See dose.

nuclear track emulsion, type A (NTA)

Film sensitive to fast neutrons made by the Eastman Kodak. The developed image has tracks 

caused by neutrons that become visible under oil immersion with about 1,000-power 

magnification.

occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

on-phantom

Exposure of a dosimeter on a phantom to simulate the dosimeter’s response when worn on a 

person.

open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). See film dosimeter.

penetrating dose

Dose from moderate to higher energy photons and neutrons that penetrate the outer layers of 

the skin. See dose.

personal dose equivalent Hp(d)

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeter 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection. 

phantom

Any structure that contains one or more tissue substitutes (any material that simulates a body 

of tissue in its interaction with ionizing radiation) and is used to simulate radiation interactions 

in the human body. Phantoms are primarily used in the calibration of in vivo counters and 

dosimeters. See slab phantom.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.
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pocket ionization chamber

Cylindrical monitoring device commonly clipped to the shirt or laboratory coat pocket to 

measure ionizing radiation. Also called pencil, pocket pencil, pencil dosimeter, and pocket 

dosimeter.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act, the percent likelihood, at the 99th percentile, that a worker 

incurred a particular cancer from occupational exposure to radiation.

proton

Basic nuclear particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element.

quality factor (QF)

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

recycled uranium (RU)

Uranium first irradiated in a reactor then recovered through chemical separation and 

purification. RU contains minor amounts of transuranic material (e.g., plutonium and

neptunium) and fission products (e.g., technetium) or uranium products (e.g., 236U) after 

purification. PGDP lists the isotopic activity ratios as:

Isotope Activity fraction

234U 0.8489

235U 0.0120

236U 0.1388

238U 0.0003 or 0.0004

(both listed)

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 
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sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

rep

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 83 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 

shallow absorbed dose

Absorbed dose at a depth of 0.07 millimeter (7 milligrams per square centimeter) in a material 

of specified geometry and composition.

shallow dose equivalent (Hs)

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeter (7 milligrams per square 

centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

sievert (Sv)

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

skin dose

See shallow dose equivalent.

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

thermoluminescent neutron dosimeter (TLND)

Thermoluminescent dosimeter for measurement of neutron dose.

tissue-equivalent proportional counter (TEPC)

Device that measures absorbed dose from neutron radiation in materials nearly equivalent to 

tissue. Analysis of the counter data determines the effective weighting factor and the dose 

equivalent for that radiation.

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 
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10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose.

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 36 of 48

SECTION TITLE PAGE

A.1 Introduction ...............................................................................................................................36

A.2 Background...............................................................................................................................36

A.3 Potential for Exposure...............................................................................................................36

A.4 Magnitude of Exposure.............................................................................................................37

A.5 Assignment of External Dose from 99Tc....................................................................................38

A.6 Assignment of Skin Contamination Dose from 99Tc..................................................................39

LIST OF TABLES 

TABLE TITLE PAGE

A-1 PGDP facilities with 99Tc exposure potential.............................................................................37

A-2 Job titles for workers with possible 99Tc exposure ....................................................................37

A.1 INTRODUCTION

This attachment provides guidance for the assignment of external dose from 99Tc for employees of 

PGDP. Due to its nonpenetrating characteristics, combined with the routine use of personal 

protective equipment (PPE) by affected employees, the dose potential from 99Tc is low. Information 

on the assignment of 99Tc dose in this attachment is based on information about work location, job 

title and description, together with shallow dose data that can be used by dose reconstructors to 

identify employees who could have been exposed to 99Tc.

A.2 BACKGROUND

Technetium-99 is present at PGDP as a contaminant from the introduction of recycled uranium into 

the cascade at various times throughout site operations. It is a long-lived fission product with a 

radiological half-life of 213,000 years and is a pure beta emitter with average and maximum energies 

of 84.6 keV and 293.6 keV, respectively. Although it is difficult to detect due to its low-energy beta 

emission, this characteristic results in minimum potential for external dose. Studies have shown that 

the outer layer of skin affords significant protection to the germinal skin layers from 99Tc, and the 

wearing of PPE such as coveralls and gloves provides further skin protection. It is in this way that the 

vast majority of radiation from 99Tc is attenuated before it can interact with the body (ORAUT 2012a).

A.3 POTENTIAL FOR EXPOSURE

Operation procedures at PGDP required special precautions for working around significant quantities 

of 99Tc, especially during plant maintenance and repairs on the upper cascade equipment. These 

included engineering controls, administrative controls, and the use of PPE by workers. The control 

measures routinely used to protect workers from exposure to uranium and its progeny provided an 

even greater protection factor for exposure to 99Tc (Saraceno 1981).

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Specific work activities that could have resulted in exposure to 99Tc included the following:

• Technetium recovery operations,

• Removal of equipment from the cascade for routine maintenance, and

• Removal and replacement of cascade equipment during the Cascade Improvement Program 

and Cascade Upgrade Program.

Table A-1 lists the facilities at PGDP with 99Tc exposure potential.

Table A-1. PGDP facilities with 99Tc exposure potential.

Facility Description

C-409 Stabilization Building

C-410 Feed Plant

C-420 Oxide Conversion Plant

C-331 Gaseous Diffusion Process Buildings

C-333 Gaseous Diffusion Process Buildings

C-335 Gaseous Diffusion Process Buildings

C-337 Gaseous Diffusion Process Buildings

C-310 Purge and Product Withdrawal Building

C-710 Analytical Laboratory

C-400 Decontamination and Cleaning Building

C-720 Maintenance Building

Employees with any of the job titles listed in Table A-2 could have had exposure to 99Tc while working 

in the facilities listed in Table A-1. The highest exposure potential would have been to maintenance 

workers in the top purge cells and to those doing change-outs of trapping media near the top purge 

cells. 

Table A-2. Job titles for workers with possible 

99Tc exposure.

Job title

Cascade worker/operator

Chemical operator

Construction trade worker

Decontamination and decommissioning worker

Feed plant operator

Maintenance mechanic

Radiological worker

A.4 MAGNITUDE OF EXPOSURE

Based on a review of the properties of 99Tc and routine controls that were in place, automatic 

assignment of 99Tc dose due to skin contamination is not warranted. Guidance for assignment of 99Tc 

skin contamination dose on a case-by-case basis is provided below. It is apparent, however, that 

external exposure to 99Tc was unlikely to be measured by dosimetry due to its low-energy electron 

characteristics. In certain cases, as described below, an annual external dose assignment from99Tc 

should be included in the dose estimate under EEOICPA. 

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Site evaluations at PGDP assessed the potential for an external exposure problem from 99Tc recovery 

operations and found that the likelihood of high exposure was low due to the following reasons (Baker 

et al. 1978): 

• Gloves were worn routinely for all operations involving the handling of containers. 

• All material was transferred remotely from point to point, with one exception. For movement

from one container to another, the transfer was done by pumping; the container was never 

dumped by hand. 

• The solutions were dilute. 

• Less than 20% of employee work time was spent at jobs with the potential to generate 99Tc 

contamination. 

Information about the magnitude of 99Tc exposure is available in Tc-99 Contamination (Swinth 2004, 

p. 3). Measured contamination exposure levels at PGDP ranging from 10,000 to 

335,849 cpm/100 cm2 was considered, which resulted in average dose rates to the skin (calculated 

using VARSKIN) that ranged from 0.212 mrem/hr on contact to 0.013 mrem/hr at a distance of 10 cm 

in air. These dose rates account for the use of coveralls with a density thickness of 28 mg/cm2

. To 

estimate the skin dose from a contamination event, a contamination level of 25,000 dpm/100 cm2

(250 dpm/cm2

) was assumed based on the action limit for 99Tc contamination on work surfaces and 

hand tools (GAT 1963). 

The dose from a contamination event is calculated as follows (Swinth 2004, p. 3): 

25,000 dpm/100 cm2 × 0.081 mrem per dpm/cm2 = 20 mrem

The assumed contamination value is greater than the average contamination level of 

13,540 cpm/100 cm2 identified by Swinth. The value of 0.081 mrem per dpm/cm2 is derived from a 

value of 1.6 × 10-3 mrem per dpm/cm2 multiplied by a residence half-time of 1.5 days. This half-time is 

assumed because 99Tc can be difficult to remove from the skin (Swinth 2004, p. 3). 

Because the low-energy 99Tc electrons would not have been detected by dosimetry, the potential 

unmeasured external electron dose can be estimated by assuming an ambient dose rate level of 

0.2 mrem/hr, a technetium-to-uranium progeny ratio of 0.4, and a 2,000-hour work year (Bassett 

1986): 

0.2 mrem/hr (maximum ambient level) × 0.4 (Tc:U progeny ratio) × 2000 hr/yr = 160 mrem/yr

Because the facilities, processes, and contaminants were similar at all three gaseous diffusion plants, 

the magnitude of exposure discussed here should be valid for PGDP.

A.5 ASSIGNMENT OF EXTERNAL DOSE FROM 99TC

The assignment of external dose due to the presence of 99Tc is warranted under certain 

circumstances for cancer sites involving the hands. Dose assignment is limited to the hands because 

the 99Tc dose rate at distances beyond 30 cm is less than 0.08 mrem/hr and drops off rapidly at 

greater distances. The following conditions must be met to assign an external dose from 99Tc: 

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1. Claimant has skin cancer on the hand(s); and

2. Claimant worked in a facility where 99Tc was present (see Table A-1); and

3. Claimant performed a job function that could have involved 99Tc exposure (see Table A-2); 

and

4. Claimant dosimetry indicates a relatively high ratio (more than 2) of shallow to deep dose 

(NIOSH 2007). 

If, and only if, all four of the above conditions are met, the dose reconstructor should: 

• Assign an external electron dose of 8 mrem/yr.

This value derives from an annual external dose of 160 mrem reduced by a protection factor of 95% 

to account for the use of PPE. The external dose should be assigned as electrons >15 keV and a 

constant distribution.

A.6 ASSIGNMENT OF SKIN CONTAMINATION DOSE FROM 99TC

Skin contamination dose due to 99Tc should be applied under certain circumstances for cancer sites 

where a documented skin contamination event occurred. The following conditions must be met to 

assign a skin contamination dose from 99Tc: 

1. Claimant has skin cancer on a potentially uncovered area of the skin; and

2. Claimant worked in a facility where 99Tc was present (see Table A-1); and

3. Claimant performed a job function that could have involved 99Tc exposure (see Table A-2); 

and

4. Claimant records indicate a contamination incident involving the area of the skin cancer site. 

If, and only if, all four of the above conditions are met, the dose reconstructor should:

• Assign a skin dose of 20 mrem per documented incident.

The skin contamination should be assigned as electrons >15 keV and a constant distribution.

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TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Purpose.....................................................................................................................................40

B.2 Background...............................................................................................................................40

B.3 Applications and Limitations .....................................................................................................41

B.4 PGDP Coworker Data Development.........................................................................................41

B.5 PGDP Coworker Annual Dose Summaries...............................................................................43

B.6 Penetrating Dose Values Based on ORAUT-OTIB-0052 Guidance for Selected 

Construction Trade Workers .....................................................................................................45

B.7 PGDP External Dosimetry Data Review...................................................................................46

LIST OF TABLES

TABLE TITLE PAGE

B-1 Missed external doses based on Section 6.3.1 of this TBD and Attachment C........................42

B-2 Annual PGDP external coworker doses modified to account for missed dose .........................45

B-3 Annual PGDP external penetrating coworker doses modified in accordance 

with ORAUT-OTIB-0052 ...........................................................................................................45

B.1 PURPOSE 

The purpose of this attachment is to provide information to enable Oak Ridge Associated Universities 

(ORAU) Team dose reconstructors to assign doses to PGDP workers who have no or limited 

monitoring data, based on site coworker data. The data in this attachment are to be used in 

conjunction with ORAUT-OTIB-0020 (ORAUT 2011a).

B.2 BACKGROUND

An analysis of external coworker dose was performed to permit dose reconstructors to complete 

certain cases for which external monitoring data are unavailable or incomplete. Cases not having 

complete monitoring data can fall into one of several categories, including:

• The worker was unmonitored and, even by today’s standards, did not need to be monitored 

(e.g., a nonradiological worker).

• The worker was unmonitored, but by today’s standards would have been monitored.

• The worker may have been monitored but the data are not available to the dose reconstructor.

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• Partial information is available but it is insufficient to facilitate a dose reconstruction.

As described in ORAUT-OTIB-0020 (ORAUT 2011a), some cases not having complete monitoring 

data can be processed based on assumptions and methodologies that do not involve coworker data. 

For example, many cases falling in the first category above can be processed by assigning ambient 

external and internal doses based on information in the relevant site profiles.

As described in Section 6.3.1 of this TBD, radiological operations at PGDP began in September 1952, 

and in 1953 the site began using dosimeter and processing technical support provided by ORNL. 

Until July 1960, dosimeters were issued to a limited number of individuals (i.e., those with the highest 

potential for exposure), and the badges were exchanged weekly. After that time, dosimeters were 

assigned to all workers who entered a controlled area, and the badges were exchanged and 

processed on a monthly or quarterly schedule. There does not appear to be any significant 

administrative practice that would jeopardize the integrity of the recorded dose of record.

B.3 APPLICATIONS AND LIMITATIONS

1. Some PGDP workers might have worked at one or more other major sites within the DOE 

complex during their employment history. Thus, the data presented herein must be used with 

caution to ensure that for clearly noncompensable cases unmonitored external doses from 

multiple site employments have been overestimated. This will typically require the availability of 

external coworker dosimetry data for all relevant sites.

2. Summary statistics based on PGDP dosimetry data in this attachment do not extend beyond 1995 

because data beyond 1997 were not available, and the data for 1996 and 1997 included too few 

data points to be considered reliable. However, the absence of these data (and the subsequent 

development of dose distributions) should not interfere with the processing of most PGDP cases 

with a lack of external dosimetry data because well before 1995 the monitoring and reporting 

practices at the site ensured that essentially all workers with a potential for external radiation 

exposure were monitored and the results are readily accessible. Coworker doses can be 

extended to later years if needed. However, the vast majority of PGDP employees with a potential 

for radiological exposure were likely to have been monitored in recent years. 

3. The data in this attachment address penetrating radiation from gamma radiation and 

nonpenetrating radiation from beta radiation. Neutron data are not presented. However, Section 

6.5.3 of this TBD should be used as the basis for assigning neutron doses, when relevant, in 

addition to the photon and beta doses assigned in accordance with this attachment.

4. External onsite ambient doses should not be included in addition to the coworker doses assigned 

in accordance with this attachment because such doses would have been included in the 

dosimetry results reported by the site, which were used as the basis for the coworker dose 

distributions presented below (ORAUT 2012b; 2006).

B.4 PGDP COWORKER DATA DEVELOPMENT

Dosimetry data for monitored PGDP workers from various sources were evaluated (see Section 7.0). 

The data selected for development of coworker doses were (1) a “history tape” containing annual data 

between 1953 and 1975 and quarterly data between 1976 and 1988, and (2) a database titled 

“OHIS_External” containing mostly quarterly data between 1989 and 1997 (although the data for 1996 

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and 1997 were excluded from consideration, as discussed above). In all cases, the reported data 

corresponded to deep doses (i.e., penetrating gamma radiation) and shallow doses (i.e., penetrating 

plus nonpenetrating radiation). 

The annual data reported between 1953 and 1975 were prorated to account for partial years of 

employment based on an analysis of the length of monitored employment associated with the data 

(see Section B.5 for further discussion). The reported quarterly data between 1976 and 1988 were 

also prorated, but with a different approach (also described in Section B.5). The data between 1989 

and 1995 included specific monitoring start and end dates, so they were prorated based on 

365 days/year. The data were prorated so coworker doses representing a full year of monitored 

employment could be derived; this permits the dose reconstructor to assign appropriate doses based 

on specific employment dates and job descriptions. 

The validity of the data selected for coworker dose development was confirmed by selecting a 

sampling of claimant dosimetry data submitted by the site as part of the EEOICPA Subtitle B program 

and comparing it with the data selected as described above. A review of annual data for 10 claimants 

with more than 150 worker-years of monitored employment at PGDP indicated excellent agreement 

between the two data sets. Specifically, a perfect match was found for more than 95% of the reported 

values. It was concluded that the data cited above are acceptable for the development of coworker 

doses for PGDP.

Adjustment for Missed Dose

According to the External Dose Reconstruction Implementation Guideline (NIOSH 2007), missed 

doses are to be assigned for dosimeter readings less than the LOD to account for the possibility that 

doses were received but not recorded by the dosimeter or reported by the site. Annual missed doses

are calculated by multiplying the number of <LOD dosimeter readings by the dosimeter LOD and 

summing the results. These values are used as the 95th percentile of a lognormal distribution for 

calculating POC; thus, in the Interactive RadioEpidemiological Program (IREP), the calculated annual 

missed doses are multiplied by 0.5 and entered in Parameter 1, and a value of 1.52 is entered in 

Parameter 2, to represent the geometric mean and geometric standard deviation, respectively.

The assignment of missed doses for monitored workers is particularly significant for PGDP claimants 

before August 1960 when workers were monitored weekly. Table B-1 lists the maximum annual 

missed dose by era and type of radiation (penetrating gamma and nonpenetrating) based on 

information in Section 6.3.1of this TBD and Attachment C.

Table B-1. Missed external doses (rem) based on Section 6.3.1 of this TBD and Attachment C.

Period

Penetrating 

LOD

Nonpenetrating 

LODa

Exchange 

frequency

Maximum annual missed dose

Penetrating Nonpenetrating

1953–1959 0.04 0.05 Weekly 2.080 2.600

1960 0.04 0.05 Variedb 1.280 1.600

1961–1980 0.04 0.05 Variedc 0.160 0.200

1981–1988 0.02 0.03 Variedd 0.080 0.120

1989–present 0.02 0.02 Quarterly 0.080 0.080

a. Attachment C provides an explanation for nonpenetrating LODs for PGDP.

b. The exchange frequency was weekly through July 1960, then became less frequent (see note c).

c. Section 6.3.1 of this TBD indicates that monthly, quarterly, or annual exchange frequencies were used during this 

period, depending on work locations and the potential for exposure. A review of the data indicates that quarterly 

exchanges were predominant; thus, quarterly exchanges have been assumed here to calculate the maximum annual 

missed dose.

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d. Section 6.3.1 of this TBD indicates that either quarterly or annual exchange frequencies were used during this period, 

depending on the potential for exposure. A review of the data indicates that quarterly exchanges were predominant; 

thus, quarterly exchanges have been assumed here to calculate the maximum annual missed dose.

Special Considerations

Certain aspects of the external dosimetry practices at PGDP as described in Section 6.3.1 of this TBD 

were considered in the analysis of the site data. These include:

• In some cases, values less than the dosimeter LODs (listed in Table B-1) were reported by the 

site. For example, values as low as a few millirem were reported even though the penetrating 

LOD was considered to be 20 or 40 mrem (depending on the era).

• As discussed above, before 1976 the data available to analyze coworker doses represent 

annual dose summaries for individual workers. Because these data include partial work years, 

the average annual doses reported tend to underestimate the average annual doses received 

by employees who worked an entire year. 

As described in Section B.5, an approach that is favorable to claimants was adopted in the 

development of coworker dose summaries; this approach is intended to account for any 

underestimate of doses to radiological workers at PGDP based on the considerations described 

above.

B.5 PGDP COWORKER ANNUAL DOSE SUMMARIES

Based on the information and approaches described above, PGDP coworker annual external 

dosimetry summaries were developed for use in the evaluation of external dose for certain claimants 

potentially exposed to workplace radiation, but with no or limited monitoring data provided by DOE. 

These summaries were developed using the following steps:

1. As described in Section B.4, for data between 1953 and 1975 the reported deep and shallow 

doses, which represented annual summary data, were modified to account for partial years of 

employment. This adjustment was made by analyzing NIOSH-OCAS Claims Tracking System 

(NOCTS) employment data for PGDP workers and adjusting the reported doses upward by an 

appropriate multiplier corresponding to the average fraction of a year an employee worked at 

the site. For example, if in a particular calendar year the average employment period for all 

PGDP employees in NOCTS was 11 months, the reported annual doses were multiplied by 

12/11, or 1.09. This permits the dose reconstructor to assign an appropriate prorated dose to 

account for partial years of employment or potential exposure.

2. For data between 1976 and 1988, the reported deep and shallow doses, which represented 

quarterly summary data, were modified to account for partial years of employment. Consistent 

with the guidelines in ORAUT-OTIB-0020 (ORAUT 2011a), doses for individuals with less than 

four quarters of data for a particular year were converted to annual doses by extrapolation

(i.e., one quarterly result was multiplied by 4; two quarterly results were multiplied by 2; and 

three quarterly results were multiplied by 1.333).

3. For data between 1989 and 1995, the reported deep and shallow doses, which represented 

primarily quarterly data, were modified to account for partial years of employment by 

multiplying the data by 365/X, where X corresponds to the number of days the employee was 

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issued a dosimeter. This information was available for this period because the monitoring start 

and end dates were included with the data.

4. Half of the maximum annual missed doses listed in Table B-1 were added to the annual doses 

from Steps 1 through 3 (with the exception of reported positive doses, in which case the 

maximum missed dose was reduced by the dose corresponding to one badge exchange 

because it is not possible that all individual badge results were zero if a positive annual dose 

was reported).

5. The 50th- and 95th-percentile annual penetrating and shallow doses were derived from the 

doses calculated in Step 4 by ranking the data into cumulative probability curves and 

extracting the 50th- and 95th-percentile doses for each year.

6. Because the reported shallow doses include penetrating and nonpenetrating radiation, the 

percentile doses pertaining to penetrating radiation identified in Step 5 were subtracted from 

the percentile doses pertaining to the reported shallow doses to derive percentile doses 

pertaining to nonpenetrating radiation. 

7. The results are listed in Table B-2. These percentile doses should be used for selected PGDP 

workers with no or limited monitoring data using the methodologies outlined in ORAUT-OTIB0020 (ORAUT 2011a). In general, the 50th-percentile dose can be used as a best estimate of 

a worker’s dose if professional judgment indicates the worker was likely to be exposed to 

intermittent low levels of external radiation. The 50th-percentile dose should not be used for 

workers who were routinely exposed. For routinely exposed workers (i.e., those who were 

expected to have been monitored), the 95thpercentile dose should be applied. For workers 

who are unlikely to have been exposed, external onsite ambient dose should be used rather 

than coworker dose.

Doses to organs affected only by penetrating radiation (e.g., organs other than the skin, breast, and 

testes) are calculated based only on the “Gamma” columns in Table B-2 combined with the 

appropriate organ DCFs (NIOSH 2007). Doses to the skin, breast, and testes (and any other cancer 

location potentially affected by nonpenetrating radiation) are determined based on both the “Gamma” 

and “Nonpenetrating” columns; gamma doses are assigned as photons with an energy range

consistent with information in this document, and nonpenetrating doses are assigned as electrons 

>15 keV with corrections applied to account for clothing attenuation or other considerations. Further 

guidance is provided in ORAUT-OTIB-0017, Technical Information Bulletin: Interpretation of 

Dosimetry Data for Assignment of Shallow Dose (ORAUT 2005). 

With the methodology described above, null values for nonpenetrating dose can occur because of the 

subtraction of the reported penetrating doses from the reported shallow doses and the method 

described above, which is favorable to claimants, to establish coworker doses based on the addition 

of potential missed doses. However, a “zero” value in Table B-2 for nonpenetrating dose will not 

result in a dose of zero to an organ such as the skin. For example, the 50th-percentile dose to the 

skin in 1989 would be assigned entirely as 0.040 rem of photons. This approach does not result in an 

underestimation of POC (which is determined by DOL) because assigning beta dose as gamma dose 

in IREP has no negative effect, because the radiation effectiveness factors are the same for >15-keV 

electrons and >250-keV photons, and are higher for 30- to 250-keV photons. 

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B.6 PENETRATING DOSE VALUES BASED ON ORAUT-OTIB-0052 GUIDANCE FOR 

SELECTED CONSTRUCTION TRADE WORKERS

Table B-3 lists penetrating dose values that have been adjusted using the guidance in Section 8.0 of 

ORAUT-OTIB-0052, Parameters to Consider When Processing Claims for Construction Trade 

Workers (ORAUT 2011b). This guidance is applicable for CTWs who meet the criteria in ORAUTOTIB-0052.

Table B-2. Annual PGDP external coworker doses modified to account for missed dose (rem).

Year

Gamma

95th%

Gamma

50th%

Nonpen

95th%

Nonpen

50th% Year

Gamma

95th%

Gamma

50th%

Nonpen

95th%

Nonpen

50th%

1953 1.656 1.128 1.729 0.701 1975 0.247 0.090 0.604 0.055

1954 2.218 1.183 4.386 0.970 1976 0.233 0.062 0.553 0.050

1955 2.344 1.067 5.574 1.048 1977 0.189 0.062 0.398 0.055

1956 2.712 1.073 4.829 1.048 1978 0.193 0.089 0.150 0.031

1957 2.224 1.072 4.511 0.580 1979 0.109 0.080 0.265 0.054

1958 2.019 1.040 4.021 0.466 1980 0.200 0.080 0.135 0.020

1959 1.900 1.083 5.148 0.694 1981 0.090 0.040 0.324 0.020

1960 1.544 0.672 3.140 0.452 1982 0.053 0.040 0.712 0.020

1961 1.048 0.134 1.647 0.036 1983 0.070 0.040 0.535 0.020

1962 1.024 0.080 1.422 0.059 1984 0.156 0.040 0.489 0.020

1963 0.868 0.080 0.818 0.037 1985 0.070 0.040 0.615 0.020

1964 0.519 0.080 0.514 0.020 1986 0.130 0.040 0.755 0.020

1965 0.243 0.080 0.194 0.020 1987 0.070 0.040 0.415 0.020

1966 0.225 0.080 0.242 0.020 1988 0.055 0.040 0.590 0.020

1967 0.236 0.091 0.343 0.025 1989 0.053 0.040 0.067 0.000

1968 0.411 0.080 0.532 0.020 1990 0.040 0.040 0.052 0.000

1969 0.541 0.080 0.989 0.020 1991 0.040 0.040 0.033 0.000

1970 0.349 0.080 0.763 0.020 1992 0.040 0.040 0.046 0.000

1971 0.558 0.080 1.039 0.020 1993 0.043 0.040 0.042 0.000

1972 0.451 0.080 1.133 0.020 1994 0.040 0.040 0.037 0.000

1973 0.407 0.080 1.254 0.020 1995 0.040 0.040 0.055 0.000

1974 0.217 0.080 0.854 0.020

Table B-3. Annual PGDP external penetrating coworker doses 

modified in accordance with ORAUT-OTIB-0052 (rem).

Year

Gamma

95th%

Gamma

50th% Year

Gamma

95th%

Gamma

50th%

1953 1.910 1.171 1975 0.322 0.102

1954 2.697 1.248 1976 0.302 0.063

1955 2.874 1.086 1977 0.240 0.063

1956 3.389 1.094 1978 0.246 0.101

1957 2.705 1.093 1979 0.128 0.080

1958 2.419 1.040 1980 0.256 0.080

1959 2.252 1.109 1981 0.114 0.040

1960 1.913 0.693 1982 0.063 0.040

1961 1.443 0.163 1983 0.086 0.040

1962 1.409 0.080 1984 0.206 0.040

1963 1.191 0.080 1985 0.086 0.040

1964 0.702 0.080 1986 0.170 0.040

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Year

Gamma

95th%

Gamma

50th% Year

Gamma

95th%

Gamma

50th%

1965 0.316 0.080 1987 0.086 0.040

1966 0.291 0.080 1988 0.065 0.040

1967 0.306 0.103 1989 0.062 0.040

1968 0.552 0.080 1990 0.044 0.040

1969 0.734 0.080 1991 0.040 0.040

1970 0.464 0.080 1992 0.040 0.040

1971 0.757 0.080 1993 0.048 0.040

1972 0.607 0.080 1994 0.040 0.040

1973 0.545 0.080 1995 0.040 0.040

1974 0.280 0.080

Source: ORAUT (2011b)

B.7 PGDP EXTERNAL DOSIMETRY DATA REVIEW

PGDP Dosimetry Data database

There are many tables in this database; they contain internal and external dosimetry data. The 

external data listings are listed below, with their descriptions.

• DRS_89_THRU_96 – External dosimetry records from 1989 to 1996.

• DRS_97_THRU_98 – External dosimetry records from 1997 to 1998.

• OHIS_EXTERNAL_DOSE – External dosimetry records from 1981 to 1997.

• OHIS_EXTREMITY_DOSE – Extremity dosimetry records from 1990 to 1995.

• OHIS_HP_SCHEDULE – Dosimetry scheduling information from 1987 to 1998.

• OHIS_JOB_HISTORY – Personnel job history information from 1986 to 1998.

• HISTORY_TAPE – External dosimetry records from 1953 to 1988.

• HIS20_EDD_CALCULATED_EXPOSURE – Calculated external dosimetry exposure records 

from 1980 to 1998.

• HIS20_EDD_INACTIVE_IRD_EXPOSURE – External dosimetry records from 1953 to 1998.

ATTACHMENT B

PGDP COWORKER DOSE ASSIGNMENT

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Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 47 of 48

TABLE OF CONTENTS

SECTION TITLE PAGE

C.1 General Information ..................................................................................................................47

C.2 Procedure .................................................................................................................................47

LIST OF TABLES

SECTION TITLE PAGE

C-1 Examples of skin dose assignments for PGDP badge readings in 1970..................................48

The guidance in this attachment is in Appendix D of ORAUT-OTIB-0017, Interpretation of Dosimetry 

Data for Assignment of Shallow Dose (ORAUT 2005), which will be withdrawn at an appropriate time.

C.1 GENERAL INFORMATION

In general, the contribution to skin dose at PGDP from low-energy photons is extremely small in 

comparison with the contribution from beta particles.

Missed doses should be calculated based on the following LODs:

• 1953 – 1980: 50 mrem for open window (OW), 40 mrem for shielded (S)

• 1981 – 1988: 30 mrem for OW, 20 mrem for S

• 1989 – present: 20 mrem for OW, 20 mrem for S

Section 6.3.1 of this TBD states an OW LOD of 120 mrem for 1953 to 1980. However, this value 

appears to be speculative when compared against the LOD values for similar dosimetry systems at 

other sites at that time. As stated in Section 6.3.1, “In 1953, PDGP began using dosimeter and 

processing technical support from ORNL ... practices were similar to those used at ORNL and other 

major sites ... ORNL has provided PGDP with dosimeters from early in the operations period through 

the present.” Based on the information, it appears that the LOD value reported is based on 

considerations involving low-energy beta emitters; however, this would significantly overestimate the 

LOD (and missed dose) when the principal source of exposure is uranium, because the dosimeters 

were calibrated using uranium slabs. Therefore, the value has been reduced in this attachment to 

10 mrem above the reported photon LOD. The dose reconstructor should consult Attachment A to 

address potential exposures to 99Tc. Table C-1 provides examples of skin dose assignments.

C.2 PROCEDURE 

Measured Dose

1. Subtract the reported S reading from the reported OW reading. This is the calculated 

nonpenetrating dose.

ATTACHMENT C

SKIN DOSE ASSIGNMENT FOR PGDP

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Document No. ORAUT-TKBS-0019-6 Revision No. 04 Effective Date: 08/24/2012 Page 48 of 48

2. Assign the calculated nonpenetrating dose as electrons >15 keV. A correction factor should 

be provided for clothing, if applicable, depending on likely clothing thickness and beta energy. 

3. Assign the reported S dose as photons, partitioned by energy according to Section 6.3.4 of 

this TBD.

4. Assign the reported neutron dose (if applicable) partitioned by energy and correct for neutron 

quality according to Section 6.4.3 of this TBD (using an organ DCF of 1).

Missed Dose

5. For a badge cycle with a zero result in the OW or S reading, or both, assign a single missed 

dose.

6. If only the OW reading was reported as zero, the missed dose assigned should be the 

appropriate OW LOD for that era (divided by 2, treated as lognormal) and considered to be 

electrons (corrected for attenuation, if applicable).

7. If only the S reading was reported as zero, the missed dose assigned should be the 

appropriate S LOD for that era (divided by 2, treated as lognormal) and considered to be 30- to 

250-keV photons.

8. If both the OW and S readings were reported as zero, the missed dose assigned should be the 

appropriate OW LOD for that era (divided by 2, treated as lognormal) and considered to be 30-

to 250-keV photons.

9. Assign missed or unmonitored neutron dose per the direction contained in Section 6.5.3 of this 

TBD. 

10. If applicable, assign unmonitored 99Tc dose (as >15-keV electrons) per the direction contained 

in Attachment A of this TBD.

Table C-1. Examples of skin dose assignments for PGDP badge readings in 1970 (assuming 

Paducah LODs, no clothing correction and no 99Tc exposure) (mrem).

OW reading S reading Measured dose assigned Missed dose assigned

50 0 50 (electrons) 40/2 = 20

(30- to 250-keV photons)

0 0 None 50/2 =25

(30- to 250-keV photons)

100 60 40 (electrons) AND 60 (photon energy per TBD) None

100 100 100 (photon energy per TBD) None

0 40 40 (photon energy per TBD) 50/2 = 25 (electrons)

ATTACHMENT C

SKIN DOSE ASSIGNMENT FOR PGDP

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